Rapid and accurate determination of the reactor power is of great importance for the control, limiting and safety systems in nuclear reactors, particularly those in utilized electric power production plants.
The use of core-external (excore) neutron detectors for determination of the reactor power in nuclear reactors is known. These measure the neutron flux outside of the reactor core, for example, from below source level when the reactor is shutdown to over 125% of full power, and may provide signals to the operators, the reactor protection system, and the reactor controls. Typical system excore neutron detectors are located in the vertical walls immediate outside the reactor vessel but inside the primary shielding. A normal signal span from source power to overpower trip is approximately ten decades (a range of 10.sup.10) and thirteen decades (a range of 10.sup.13) of neutron flux information to include margins at both ends of the operating scale to allow for variations in source strength. These neutron-flux measurement signals are nearly prompt (instantaneous) but in large reactor cores are not always proportional to the reactor power output, particularly with load transients. Moreover, the measured signals of core-external neutron detectors must be calibrated either manually for the reactor powers determined from the reactor heat balance calculated by the plant computer or automatically for the reactor powers determined from the warm-up range of the coolant in the reactor core.
In pressurized water reactors, the nuclear reactor is cooled by water as a primary coolant under considerable pressure so that the average enthalpy of the water leaving the reactor is less than the enthalpy at saturation temperature. The high pressure primary coolant is conducted to steam generators and steam is produced on a low pressure or secondary side from feedwater which enters the steam generators. A heat balance may be performed on either the primary or secondary side.
The automatic calibration of the neutron-flux measurement signal with the aid of the reactor power determined from the warm-up range of the coolant in the primary loop normally has been effected by means of a closed-loop control system. The drawback of this is that with fast power changes calibration cannot follow the reactor power quickly enough. This in turn means that with fast load changes the automatic calibration system must be disabled as it might result in nonconservative power values.
While determination of the reactor power from a heat balance on the primary or secondary side is considerably more accurate, for physical reasons, than its determination from core-external instrumentation, this thermal power measurement system is relatively slow, compared to the prompt indication of core-external neutron detectors, and the signal output thereof, under transient conditions, lags mainly because of the transit time of the coolant between the temperature measuring points at the inlet to and outlet from the core.
Varying approaches, differing from that of the invention described hereafter, have been suggested in the background art to exploit the advantages while avoiding the limitations of power measurements dependent on neutron flux and thermal parameters. U.S. Pat. No. 3,752,735, for example, teaches using the difference between a neutron flux power signal and a thermal power signal (a signal based on coolant thermal parameters) to adjust neutron flux power signals. U.S. Pat. No. 3,356,577 discloses a similar technique.